Dong-Seong Sohn

Ulsan National Institute of Science and Technology

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Yang-Hyun Koo

Korea Atomic Energy Research Institute

Byung-Ho Lee

Pohang University of Science and Technology

Jin-Sik Cheon

Korea Atomic Energy Research Institute

Byung‐Ho Lee

Kyungpook National University

Je-Yong Oh

Korea Atomic Energy Research Institute

Gwan Yoon Jeong

Korea Atomic Energy Research Institute

Young‐Woo Lee

Soonchunhyang University

Cheol-Min Lee

Seokyeong University

Jae-Yong Oh

Chunnam Techno University

Yeon Soo Kim

Argonne National Laboratory

Cooperative Institutions

Korea Atomic Energy Research Institute

103

Seoul National University

Ulsan National Institute of Science and Technology

Korea Advanced Institute of Science and Technology

Sungkyunkwan University

Massachusetts Institute of Technology

Pusan National University

Chungnam National University

Korea University

Yonsei University

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Modeling of the Rim Effect on Thermal Behavior of High-Burnup UO₂Fuel

Nuclear Technology (1999)

Byung-Ho Lee Yang-Hyun Koo Dong-Seong Sohn

A model for rim porosity that takes into account the effect of overpressurization on rim pores is proposed for high-burnup UO2 fuel. It is based on the assumption that all the fission gases produced are retained in rim pores, and the threshold pellet average burnup required for the formation of the rim region is 40 MWd/kg U. In addition, a thermal conductivity correlation is proposed that uses the rim porosity model developed. This correlation for the rim region considers both degradation of thermal conductivity with burnup across the fuel pellet and additional degradation at the pellet rim due to very high porosity. To calculate the temperature profile across the fuel pellet where the rim region is formed, the present correlation for the rim region is combined with the HALDEN, MATPRO, and SIMFUEL correlations for thermal conductivity for the fuel interior region where the rim feature does not exist. Application of the present correlation to the measured HALDEN fuel centerline temperature (Nuclear Energy Agency public database IFA-562) shows that good agreement between the calculated and measured fuel centerline temperature is obtained when the present correlation is combined with HALDEN thermal conductivity. On the other hand, when it is combined with SIMFUEL thermal conductivity, which does not consider the effect on thermal conductivity of fission gases and other volatile fission products, lower centerline temperature is obtained due to the characteristics of the SIMFUEL.

Burnup

Pellet

Nuclear Fuel

10.13182/nt99-a2991

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Citations (4)

Thermophysical properties of sintered aluminum-silicon

Journal of Alloys and Compounds (2018)

Tae Won Cho Yeon Soo Kim Dong-Seong Sohn

Particle (ecology)

10.1016/j.jallcom.2018.03.368

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Citations (6)

Modeling and parametric studies of the effect of inhomogeneity on fission gas release in LWR MOX fuel

Annals of Nuclear Energy (2002)

Yang-Hyun Koo Byung-Ho Lee Jin-Sik Cheon Dong-Seong Sohn

Fuel element failure

Sodium-cooled fast reactor

10.1016/s0306-4549(01)00035-4

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Citations (17)

Enhanced high-temperature oxidation resistance of a zirconium alloy cladding by high-temperature preformed oxide on the cladding

Corrosion Science (2017)

Cheol-Min Lee Dong-Seong Sohn

Cladding (metalworking)

Zirconium alloy

High-Temperature Corrosion

10.1016/j.corsci.2017.11.019

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Citations (37)

Modeling of creep behavior of Zircaloy-4 by considering metallurgical effect

Annals of Nuclear Energy (2002)

Byung‐Ho Lee Yang-Hyun Koo Jin-Sik Cheon Dong-Seong Sohn

Rod

Zirconium alloy

Cladding (metalworking)

10.1016/s0306-4549(01)00030-5

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Citations (6)

Thermal conductivities of actinides (U, Pu, Np, Cm, Am) and uranium-alloys (U–Zr, U–Pu–Zr and U–Pu–TRU–Zr)

Journal of Nuclear Materials (2013)

Yeon Soo Kim Tae Won Cho Dong-Seong Sohn

10.1016/j.jnucmat.2013.11.018

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Citations (41)

Cosmos: A computer code to analyze LWR UO2 and MOX fuel up to high burnup

Annals of Nuclear Energy (1999)

Yang-Hyun Koo Byung-Ho Lee Dong-Seong Sohn

Burnup

Nuclear Fuel

Transient (computer programming)

10.1016/s0306-4549(98)00033-4

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Citations (30)

Corrosion and mechanical properties of hot-rolled 0.5%Gd-0.8%B-stainless steels in a simulated nuclear waste treatment solution

Nuclear Engineering and Technology (2018)

Moo Young Jung Youl Baik Yong Choi Dong-Seong Sohn

Corrosion and mechanical behavior of the hot-rolled 0.5%Gd-0.8%B-stainless steel to develop a spent nuclear fuel storage material was studied in a simulated nuclear waste treatment condition with rolling condition. The austenite and ferrite phases of the 0.5%Gd-0.8%B-stainless steels are about 88:12. The average austenite and ferrite grain size of the plane normal to rolling, transverse and normal directions of the hot rolled specimens are about 5.08, 8.94, 19.35, 23.29, 26.00 and 18.11 [μm], respectively. The average micro-hardness of the as-cast specimen is 200.4 Hv, whereas, that of the hot-rolled specimen are 220.1, 204.7 and 203.5 [Hv] for the plane normal to RD, TD and ND, respectively. The UTS, YS and elongation of the as-cast and the hot-rolled specimen are 699, 484 [MPa], 34.0%, and 654, 432 [MPa] and 33.3%, respectively. The passivity was observed both for the as-cast and the hot rolled specimens in a simulated nuclear waste solution. The corrosion potential and corrosion rate of the as-casted specimens are −343 [mVSHE] and 3.26×10−7 [A/cm2], whereas, those of the hot rolled specimens with normal to ND, RD and TD are −630, −512 and −620 [mVSHE] and 6.12×10−7, 1.04×10−6 and 6.92×10−7 [A/cm2], respectively. Corrosion tends to occur preferentially Cr and B rich area.

Hot rolled

Elongation

Austenitic stainless steel

10.1016/j.net.2018.08.017

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Citations (10)

Analysis of fission gas release and gaseous swelling in UO2 fuel under the effect of external restraint

Journal of Nuclear Materials (2000)

Yang-Hyun Koo Byung‐Ho Lee Dong-Seong Sohn

Burnup

Pellet

Isothermal process

10.1016/s0022-3115(00)00034-9

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Citations (23)

Irradiation Test of MOX Fuel Rods Fabricated by Attrition-Milling and Analysis of In-Pile Data with COSMOS Code

Nuclear Technology (2010)

Byung-Ho Lee Yang-Hyun Koo Han‐Soo Kim Jae-Yong Oh Young‐Woo Lee

Attrition-milling technology for fabricating mixed oxide (MOX) fuel was developed to mix the plutonium in UO2 fuels as homogeneously as possible. The fabricated MOX fuels were instrumented with temperature and pressure gauges that enabled one to measure the fuel temperature and rod internal pressure online. An irradiation test in the Halden reactor was performed to investigate the in-pile behavior of the fabricated MOX fuel. The irradiation of 1020 effective full-power days was successfully accomplished with good integrity of the test fuel rods. The rod average burnup reached ~50 MWd/kg HM, and the measured fuel centerline temperature was ~1000°C for the MOX fuels. A significant fission gas release was observed due to the high power level. The online measured in-pile performance data of the two attrition-milled MOX fuel rods were analyzed and compared with the fuel performance code COSMOS. COSMOS simulated the fuel centerline temperature and rod internal pressure for both MOX fuel rods. The analysis by COSMOS showed good agreement with the online measured in-pile behavior of MOX fuel.

Rod

Burnup

Nuclear Fuel

10.13182/nt10-a10933

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Citations (0)